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Title: Pre-conceptual Development and characterization of an extruded graphite composite fuel for the TREAT Reactor

In an effort to explore fuel systems that are more robust under accident scenarios, the DOE-NE has identified the need to resume transient testing. The Transient Reactor Test (TREAT) facility has been identified as the preferred option for the resumption of transient testing of nuclear fuel in the United States. In parallel, NNSA’s Global Threat Reduction Initiative (GTRI) Convert program is exploring the needs to replace the existing highly enriched uranium (HEU) core with low enriched uranium (LEU) core. In order to construct a new LEU core, materials and fabrication processes similar to those used in the initial core fabrication must be identified, developed and characterized. In this research, graphite matrix fuel blocks were extruded and materials properties of were measured. Initially the extrusion process followed the historic route; however, the project was expanded to explore methods to increase the graphite content of the fuel blocks and explore modern resins. Materials properties relevant to fuel performance including density, heat capacity and thermal diffusivity were measured. The relationship between process defects and materials properties will be discussed.
Authors:
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Publication Date:
OSTI Identifier:
1186757
Report Number(s):
INL/CON-14-32449
DOE Contract Number:
DE-AC07-05ID14517
Resource Type:
Conference
Resource Relation:
Conference: TMS2015,• Walt Disney World, Orlando, Florida, USA,03/15/2015,03/19/2015
Research Org:
Idaho National Laboratory (INL)
Sponsoring Org:
DOE - NE
Country of Publication:
United States
Language:
English
Subject:
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS LEU Fuel, Graphite Composite, TREAT